The present invention relates to a dispersion strengthened ferritic steel for high temperature structural use which has excellent high temperature strength, ductility and toughness, and a reduced strength anisotropy.
The dispersion strengthened ferritic steel of the present invention is not only suitable as a core member of a nuclear reactor, particularly a fast breeder reactor but also can be advantageously utilized as a high temperature member of structures of equipment, e.g., piping members of a cooling system and boiler tubes, used under severe temperature and service conditions.
An example of the high temperature member, i.e., a material used as a core member of a nuclear reactor, particularly a fast breeder reactor is required to have various characteristics such as high temperature strength, compatibility with sodium, resistance to neutron radiation, workability, weldability, and interaction between the member and nuclear fuel. In particular, the high temperature strength and the resistance to neutron radiation are important factors in determining the service life.
Although an austenitic stainless steel, such as SUS 304 or 316, has hitherto been used as a reactor core member, it is known that this material has limited resistance to fast neutron, such as swelling resistance and irradiation creep characteristics, and therefore is unsuitable for prolonging the service life of nuclear fuel.
On the other hand, although the ferritic steel exhibits irradiation resistance far superior to that of the austenitic stainless steel, it is disadvantageously low in the high temperature strength. Dispersion strengthening with fine oxide particles is known for long as a method of improving the high temperature strength. Examples of the ferritic steel produced by this method are disclosed in a prior art reference, U.S. Pat. No. 4075010 entitled "Dispersion-strengthened ferritic alloy for use in liquid-metal fast breeder reactors (LMFBRS)". (The alloy disclosed in the U.S. Patent is hereinafter referred to as "the prior art alloy".)
Although the prior art alloy has high strength, it has low ductility and a ductile-brittle transient temperature as high as about 20.degree. C, i.e., exhibits a very low impact value at room temperature, which brings about cracking even when the percentage cold rolling is as low as about ten-odd %. Therefore, it is difficult to economically produce from the prior art alloy core members of a fast breeder reactor, e.g., thin-wall pipes such as a fuel chadding tube or a wrapper tube which should be prepared with high dimensional accuracy. Further, the prior art alloy is a low ductility material which causes the cracks to be very easily propagated at a service temperature of the fast breeder reactor, i.e., 350.degree. to 700.degree. C. In other words, this alloy exhibits no advantages inherent in the dispersion strengthened material.
The dispersion strengthened ferritic steel has a problem of the so-called anisotropy of the strength that the strength in the direction perpendicular to the direction of working is 1/2 to 1/3 of the strength in the direction parallel to the direction of the working due to elongation of grains in the direction of working.